This invention relates to a method for immobilizing radioactive wastes for permanent disposal. More particularly, the invention relates to a method of immobilizing mixed waste chloride salts containing radionuclides and other hazardous materials for permanent disposal.
The recovery of fissionable materials such as uranium and plutonium from spent nuclear reactor fuels can be carried out by electrorefining method using electrochemical cells of the type described in U.S. Pat. Nos. 4,596,647 and 2,951,793, as well as U.S. Pat. No. 4,880,506. It is the electrorefining method which is being developed for the reprocessing of integral fast reactor (IFR) fuel. In a typical electrorefining cell, an electrolyte consisting of a molten eutectic salt mixture such as KCl and LiCl is used to transport the metal or metals to be purified between electrode solutions. When used to reprocess spent nuclear reactor fuels, the salt mixture becomes contaminated with radionuclides, such as cesium.sup.-137 and strontium.sup.-90, hazardous metals such as barium and other species such as sodium and iodine.sup.-129 and eventually is no longer suitable for use in the electrorefining cell.
Ideally the salt would be decontaminated by removing the heat producing radionuclides, primarily cesium and strontium, and any other metals, e.g. sodium, which could potentially interfere in the operation of the electrorefiner and the purified salt would be recycled back to the electrorefiner. However, the separation of cesium and strontium chloride from the salt is difficult, and since they are large heat producers it would be necessary to dilute them in another matrix material and/or cool them before they could be stored. It is therefore more practical to dispose of the cesium and strontium and any other radionuclides and toxic metal chlorides and iodides along with a portion of the salt matrix. The waste salt containing the cesium and strontium is a high level waste (HLW), and as such must be disposed of in the geologic repository for HLW. This requires that the waste form be leach resistant to prevent an uncontrolled release of the radionuclides and other hazardous chemicals such as barium into the groundwater. Since waste salts are chlorides and are very water soluble, a method for encapsulating and immobilizing the waste salt must be identified.
One problem with developing a waste storage medium is that the waste salt consists primarily of chloride salts of the alkali metals and as such is not readily amenable to treatment using procedures and techniques developed for immobilizing the cesium and strontium in other nuclear waste streams. For instance, the chloride salts cannot be added directly to glass-forming compounds and processed to yield a leach-resistant glass since glasses containing halides ions are relatively water soluble. Therefore, for immobilization in a glass matrix the waste chloride salts must be converted into oxides or other chemical forms compatible with the glass-making process. However, conversion processes are expensive and time-consuming and raise environmental concerns about the off-gases produced by the processes. A mortar matrix has also been considered as a possible waste form for the waste chloride salt. A special mortar was developed to incorporate lithium, potassium, cesium and strontium chloride salts into its structure and thereby immobilize them. However, when irradiated, the water in the mortar was radiolyzed and large quantities of hydrogen gas were generated.
A new matrix for immobilizing waste chloride salts is therefore needed. Zeolites which can be treated with molten salts are potential candidates because of their sorption and ion exchange properties. When some zeolites are treated with molten salts, salt molecules penetrate the cavities and channels of the zeolite and are then said to be occluded. Occluded molecules provide a transfer medium for ion exchange between the cations in the zeolite and those in the bulk salt. A zeolite which has a high selectivity for cesium, strontium and barium would be a promising candidate for an immobilization matrix.
The ion exchange and sorption properties of zeolites in molten salts has been investigated in several studies. Most of the studies, though involved nitrate salts, not chloride salts.
The ion exchange properties of several zeolites have been investigated in molten nitrate salt solutions [C. M. Callahan, J. Inorg. Nucl. Chem., 28, 2743 (1966)]. Callahan reported the distribution coefficients (concentration in the zeolite phase/concentration in the salt phase) for sodium, calcium, potassium, rubidium, cesium and barium between chabazite and three solvents salts, LiNO.sub.3, NaNO.sub.3 and KNO.sub.3. In each solvent salt one cation was preferentially sorbed, i.e., the distribution coefficient for the preferred cation could be as much as 100 times greater than that for the other cations. This study also showed that ion exchange was minimal when KNO.sub.3 was the solvent salt. The distribution coefficients of all the solute cations in KNO.sub.3 were very small, varying from &lt;1 to 8.7, far less than those in LiNO.sub.3 and NaNO.sub.3 where distribution coefficients could exceed 100.
Ion exchange was investigated between salt occluded sodium A zeolite, Na.sub.12 [(AlO.sub.2).sub.12 (SiO.sub.2).sub.12 ].10NaNO.sub.3, and solutions of silver, lithium, potassium, cesium, thallium, calcium, and strontium in molten NaNO.sub.3. It was found that silver and `presumably` lithium could be completely exchanged with all 22 sodium ions but that exchange with the larger ions was limited.
The occlusion of lithium, sodium and potassium nitrate in the respective forms of zeolite A was also studied, i.e., LiA, NaA and KA. The results were that LiNO.sub.3 and NaNO.sub.3 were readily occluded but KNO.sub.3 was either not occluded at all or occluded to a very limited extent.
Another study by Susic et al [J. Inorg. Nucl. Chem., 33, 2667 (1971)] investigated the salt occlusion and ion exchange properties of zeolite A in molten alkali halides, sulfates and nitrates. This study was primarily concerned with measuring the amount of occluded salt for the various melts. It was reported that 3.5 meq of chloride per gram was occluded when LiCl-KCl eutectic salt was equilibrated with the lithium form of zeolite A (LiA). The report indicated that ion exchange between lithium and potassium did not occur and inferred that LiCl, but not KCl, was occluded by LiA.
None of the studies cited above were directly applicable to the problem of predicting whether any zeolite would sorb cesium, strontium and barium from a complex salt mixture such as IFR waste salt (one consisting primarily of LiCl and KCl with smaller amounts of NaCl) and retain them so that the zeolite with the occluded salt would act as an immobilization matrix for the cesium, strontium and barium as well as for the matrix (lithium, potassium and sodium) chloride salts. In general the prior art teaches that steric factors are very important in ion exchange and salt occlusion in molten salt-zeolite systems. The presence of an excess quantity of a large ion such as potassium salt will cause the exclusion of other large ions such as cesium and barium. Occlusion and hence the desired ion exchange with cesium, strontium and barium will not occur.